STATUS OF IRSN LEVEL 2 PSA (PWR 900) General objectives Content of the study Level 1 to Level 2 Interface Quantification of physical phenomena with uncertainties in APET A model for containment leakage through containment penetrations Radioactive releases model KANT : a quantification software for level 2 PSA CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

General objectives A level 2 PSA for French 900 MW PWR to contribute to reactor safety level assessment, to estimate the benefits of accident management procedures, to provide quantitative elements about advantages of any reactor design or operation modifications, to acquire quantitative knowledge for emergency management teams, to help in definition of RD programs in the severe accident field learning from detailed studies are also extended to other French Plants CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Steps 2000 - version (1.0) based on IRSN level 1 PSA published in 1990 – power states of reactor 2003 – version (1.1) - revision of 1.0 - power states of reactor 2004 – version 2.0 - updated level 1 PSA – response surfaces method for uncertainties assessment - hydrogen recombiners 2005 – version 2.1 – shutdown states of reactor CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Content General methodology initially based on NUREG 1150 Binning of level 1 PSA sequences in PDS Representation of important severe accident events in an APET Binning of level 2 PSA into Release Categories Assessment of radioactive releases for each release category Uncertainties assessment by Monte-Carlo method CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

A detailed interface between level 1 to level 2 PSA 20 interfaces variables serve to define the Plant Damage States and concern initiator event, system and containment state, residual power, activation of emergency plan. PT – RCS break size SF – Component cooling or essential service water systems PL – RCS break localization AP – Water makeup to RCS availability RT – SGTR number BA – Safety injection water tank VL – V-LOCA SE – Secondary system break AS – CHRS availability SO – Pressurizer safety valve availability BP – Low pressure safety injection availability IE – Containment isolation HP – High pressure safety injection availability CR – Core criticity GV – SG availability PR – Residual power LC – Electrical board availability (low voltage) PU – Emergency plan LH – Electrical board availability (high voltage) RS – Electrical network availability CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

A detailed interface between level 1 to level 2 PSA A high level of description of system states Examples AS variables values 1 = CHRS available and in service 2 = CHRS available and not in service 3 = CHRS not available, failure occurred at demand 4 = CHRS not available, failure occurred in function – not contaminated 5 = CHRS not available, failure occurred in function – contaminated CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

A detailed interface between level 1 to level 2 PSA 150 Plant Damage States have been defined for power states. A representative thermal-hydraulics transient is defined for each PDS   Number of PDS Number of thermal-hydraulics transients LOCA (large break) 17 9 LOCA (medium break) 24 14 LOCA (small break) 8 LOCA (very small break) 10 SGTR 20 15 Secondary break 13 Loss of heat sink Loss of steam generator water injection Total loss of electrical power 12 6 CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

A detailed interface between level 1 to level 2 PSA Thermal-hydraulics transient are calculated with the SCAR version of the simulator SIPA 2 (that includes CATHARE 2). Advantages of this approach : to obtain a better evaluation of accident kinetics and delays before releases, to consolidate level 1 PSA assumptions, to define more precise conditions for severe acc. Phenomena, to provide a large panel of « best-estimated » transients for use in other context (accident management team, safety analysis) CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

APET – Quantification of physical phenomena with uncertainties The different physical phenomena are organized in « physical models » : each physical model represents a set of physical phenomena that are tightly coupled ; 2 separated models are linked by a limited numbers of variables transmitted by the APET CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Physical models of APET Level 1 PSA Plant Damage State Before Core degradation During Core degradation Vessel Rupture Corium-Concrete Interaction Before core degradation I- SGTR During Core Degradationn Advanced core degradatio Combustion H2 In-vessel steam explosion Direct Containt Heating Containment mechanical behavior Corium concrete interaction Ex-vessel s.e. CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Physical models of APET Codes Construction of physical model based on results obtained by validated codes calculations. Expert’s judgments are used for result interpretation or when direct code calculations are note possible CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Physical models of APET Two methods are employed METHODE 1 : RESPONSE SURFACES Downstream variables values = F(upstream variables values) (Details provided in second workshop presentation) METHODE 2 : GRID OF RESULTS For core degradation progression strong scenario effects and discontinuities have to be taken into account (valve opening, RCS cooling by SG, RCS water injection …) Construction of response surfaces would be a very difficult task Grid of result approach is used CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Physical models of APET Example of Core degradation STEP 1 : DEFINITION OF CALCULATIONS STEP 2 : CONSTITUTION OF A RESULT GRID Core degradation transient without actions recommended by severe accident management guides PDS TH-system transient Core degradation transient with actions recommended by severe accident management guides Transient N° Identification variables values DCD downstream (results) variables values CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Physical models of APET Example of Core Degradation STEP 3 : RESULT GRID IN THE APET ONE SCENARIO DEPENDS ON SYSTEM AVAILIBILITY, HUMAN ACTIONS, RESIDUAL POWER … A SELECTION TREE SELECTS THE MOST REPRESENTATIVE TRANSIENT IN THE RESULTS GRID THE DOWNSTREAM VALUES ARE EXTRACTED FROM THE RESULTS GRID FOR THE REPRESENTATIVE TRANSIENT CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Leakage through containment penetrations « b mode » A specific method has been developed to take into account pre-existing leakage or isolation failure during the accident A specific software, BETAPROB has been developped A model is constructed : System description (hydraulics components, valves, pumps, sumps, rooms of auxiliary building and ventilation/filtration level) Failure probabilities (l, failure in operation, g, failure on demand) Severe (100 % section) and non severe (1% section) are distinguished) CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Leakage through containment penetrations APET Model For each system configuration, BETAPROB calculates all the possible leakage paths and proposes a classification of leakage paths as a function of Nature of release source (liquid from RCS or gaseous from containment atmosphere) Transfer mode to environment in function of ventilation systems and filtration Leakage section In the APET, for each systems configurations are calculated Probabilities of leak categories in term of leakage section Probabilities of leak categories in term of filtration efficiency CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

The radioactive releases calculation model A simplified model has been developed for level 2 PSA. Each level 2 sequence is characterized by « APF » variables that give information on accident progression and containment failure. The model can calculate radiaoactive releases as a time function of time for each combination of APF variables. Uncertainties have been taken into account for most influent parameters. CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

The radioactive releases calculation model Fission product emission Noble Gases Volatil molecular iodine Progressive Aerosol Emission Melt - corium 1100 °C First corium flow Vessel Break CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Fission products behavior in containment Containment atmosphere composition Aerosol mass in suspension depends on : emission, energetic phenomena in RCS (steam explosion) or in containment (Combustion), natural deposition, spray system (CSHRS) efficiency and containment leakage Molecular iodine depends on : emission, painting adsorption, spray system (CSHRS) efficiency and containment leakage Organic iodine depends on : adsorbed molecular iodine to organic iodine and containment leakage Noble gases depends on : emission and containment leakage Radioactive releases depend on Containment leakage size (mass flow), Containment atmosphere composition, Aerosol filtration and iodine retention, Activity as a function of delay after SCRAM CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

The radioactive releases calculation model Graphical interface A graphical interface allows interactive calculation in function of APF variables values CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

KANT A software for level 2 PSA quantification A specific software, able to take into account the specifities of the IRSN methodologies has been developed. The software is linked with the releases model Operational for Windows operating system (C++, MFC, Access) 3 main modules : APET development (subtrees, specific language for model) APET quantification (Monte-Carlo method) Results vizualization CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

KANT Example of results vizualization CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

KANT Perspectives Future Improvements Extension of functionalities in terms of results presentation Identification and quantification of early radioactive releases Graphical presentation of the APET A convivial interface to give access to main results CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

Conclusions A detailed level 2PSA for French 900 MW is performed by IRSN with some specifities Systematic use of validated codes Original models (containment leakage, human factor) Detailed interface and large transient calculation A specific software, KANT, operational since 1998, with a development program Future 2004 – Analysis of French Utility approach for level 2 PSA 2004 – Version 2.0 for power states of reactor (recombiner, …) 2005 – Version 2.1 for shutdown states of reactor 2006 ? Improvement of methods (dynamic fiability ?, interface ?), Other plant application (?) CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004